Generic Aging Management Programs for License Renewal of PWR Reactor Coolant System Components

2002 ◽  
Author(s):  
V. N. Shah ◽  
Y. Y. Liu
Keyword(s):  
Extended Period ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Aging Effects ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Water Reactors ◽  
Regulatory Commission ◽  
License Renewal ◽  
Coolant System

The paper reviews the existing aging management programs (AMPs) for the reactor coolant system (RCS) components in pressurized water reactors (PWRs), including the reactor pressure vessel and internals, the reactor coolant system and connected lines, pressurizer, reactor coolant pumps, valves, and steam generators. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process. These AMPs include both generic and plant-specific programs. The generic AMPs are acceptable for managing aging effects during an extended period of operation and do not require further evaluation; the plant-specific AMPs require further evaluation. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.

scholarly journals Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

2002 ◽  
Author(s):  
V. N. Shah ◽  
Y. Y. Liu
Keyword(s):  
Operating Experience ◽  
Extended Period ◽  
Aging Effect ◽  
Lessons Learned ◽  
Monitoring And Control ◽  
Aging Effects ◽  
Management Activity ◽  
Reactor Coolant ◽  
License Renewal ◽  
Coolant System

The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of the AMPs, which do not require further evaluation, do need enhancements to allow for an extended period of reactor operation. The paper concludes that the GALL report systematically evaluates the generic AMPs for their effectiveness during an extended period of operation and provides a technical basis for their acceptance. The use of the GALL report should facilitate both preparation of a license renewal application and timely and uniform review by the NRC staff.

Aging Management of Reactor Coolant System Mechanical Components in Pressurized Water Reactors for License Renewal

2002 ◽  
Author(s):  
M. Subudhi ◽  
R. Morante ◽  
A. D. Lee
Keyword(s):  
Reactor Vessel ◽  
Pressurized Water Reactors ◽  
Pressurized Water ◽  
Reactor Coolant ◽  
Management Review ◽  
Mechanical Components ◽  
Water Reactors ◽  
Effects Of Aging ◽  
License Renewal ◽  
Coolant System

The reactor coolant system (RCS) mechanical components in pressurized water reactors (PWRs) that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer, steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions, determination of the effects of aging on their intended safety functions, and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. This paper presents a number of generic issues, including the time-limited aging analyses, associated with RCS components that require further review by the staff.

License Renewal Demonstration Project

2002 ◽  
Author(s):  
Barry J. Elliot ◽  
Jerry Dozier
Keyword(s):  
United States ◽  
State Of The Art ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Management Program ◽  
Lessons Learned ◽  
Demonstration Project ◽  
Aging Effects ◽  
Regulatory Commission ◽  
License Renewal

Generic Aging Lessons Learned (GALL) report, License Renewal Standard Review Plan (SRP-LR), and regulatory guide were issued by the United States Nuclear Regulatory Commission (NRC) in June 2001. The intent of these documents was to provide the technical and process basis that will lead to a more effective, efficient and predictable license renewal process for industry and the NRC. The GALL report provides the aging effects on components and structures, identifies the relevant existing plant programs, and evaluates the program attributes to manage aging effects for License Renewal. The GALL report also identifies when existing plant programs would require further evaluation for License Renewal. The SRP-LR allows the applicant to reference the GALL report to demonstrate that the programs at the applicant’s facility correspond to those reviewed and approved in the GALL report. Programs that correspond to those in the GALL report will not need further detailed review by the staff. Implementation of the aging management program are verified as part of the license renewal inspection program. The GALL report identifies one acceptable way of demonstrating that components and structures have adequate aging management programs. However, applicants may propose alternatives to the programs identified in GALL. During the license renewal review, the NRC primarily focuses on areas where existing programs should be augmented or new programs developed for License Renewal. This paper will provide an overview of these documents and some of the lessons learned during a demonstration project in the application of the new guidance. This topic will be of interest to the U.S. participants considering License Renewal and desiring to know state-of-the-art information about License Renewal in the United States.

Application of Common Basis Stress Evaluation Methodology for Environmentally Assisted Fatigue for a Pressurized Water Reactor

2013 ◽  
Author(s):  
William F. Weitze ◽  
Timothy D. Gilman ◽  
Lora Drenth
Keyword(s):  
Water Environment ◽  
Nuclear Regulatory Commission ◽  
Lessons Learned ◽  
Evaluation Methodology ◽  
Pressurized Water ◽  
Stress Evaluation ◽  
Common Basis ◽  
Consistent Manner ◽  
Regulatory Commission ◽  
License Renewal

United States (US) Nuclear Regulatory Commission (NRC) report NUREG-1801, the Generic Aging Lessons Learned (GALL) Report [1], identifies acceptable aging management programs, including programs for fatigue and cyclic operation, and is used by the NRC to evaluate license renewal applications. This includes fatigue usage analyses that account for reduced fatigue life for components in a reactor water environment. Originally, it was considered acceptable for the purpose of license renewal to evaluate only the sample locations identified in NUREG/CR-6260. Recently, however, the NRC staff has been requesting license renewal applicants to demonstrate that the locations identified in NUREG/CR-6260 are the limiting locations for environmentally assisted fatigue (EAF), such that EAF evaluations are limited to the NUREG/CR-6260 locations. Any locations not bounded by the NUREG/CR-6260 locations would then be specifically addressed. A methodology has been developed for EPRI, called common basis stress evaluation (CBSE), to perform simplified stress and fatigue usage analyses in a consistent manner such that many locations can be compared for fatigue usage with and without the use of EAF, even if prior analyses of these locations were not consistent in level of detail, and even if some locations have never been evaluated for fatigue usage. This paper presents the first application of this methodology.

scholarly journals Analysis of the Technological Evolution of Materials Requirements Included in Reactor Pressure Vessel Manufacturing Codes

2021 ◽  
Vol 13 (10) ◽  
pp. 5498
Author(s):  
Alvaro Rodríguez-Prieto ◽  
Mariaenrica Frigione ◽  
John Kickhofel ◽  
Ana M. Camacho
Keyword(s):  
Sustainable Development ◽  
Pressure Vessel ◽  
Nuclear Regulatory Commission ◽  
Green Energy ◽  
Reactor Pressure Vessel ◽  
Technical Literature ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission ◽  
Reactor Pressure

The growth of green energy technologies within the frame of the 7th Sustainable Development Goal (SDG) along with the concern about climatic changes make nuclear energy an attractive choice for many countries to ensure energy security and sustainable development as well as to actively address environmental issues. Unlike nuclear equipment (immovable goods), which are often well-catalogued and analyzed, the design and manufacturing codes and their standardized materials specifications can be considered movable and intangible goods that have not been thoroughly studied based on a detailed evaluation of the scientific and technical literature on the reactor pressure vessel (RPV) materials behavior. The aim of this work is the analysis of historical advances in materials properties research and associated standardized design codes requirements. The analysis, based on the consolidated U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.99 Rev.2 model, enables determination of the best materials options, corresponding to some of the most widely used material specifications such as WWER 15Kh2MFAA (used from the 1970s and 1980s; already in operation), ASME SA-533 Grade B Cl.1 (used in pressurized water reactor-PWR 2nd–4th; already in operation), DIN 20MnMoNi55 and DIN 22NiMoCr37 (used in PWR 2nd–4th) as well as ASTM A-336 Grade F22V (current designs). Consequently, in view of the results obtained, it can be concluded that the best options correspond to recently developed or well-established specifications used in the design of pressurized water reactors. These assessments endorse the fact that nuclear technology is continually improving, with safety being its fundamental pillar. In the future, further research related to the technical heritage from the evolution of materials requirements for other clean and sustainable power generation technologies will be performed.

A Risk-Informed Methodology for ASME Section XI, Appendix G

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Ronald Gamble ◽  
William Server ◽  
Bruce Bishop ◽  
Nathan Palm ◽  
Carol Heinecke
Keyword(s):  
Nuclear Regulatory Commission ◽  
The United States ◽  
Service Level ◽  
Pressure Vessels ◽  
Operational Flexibility ◽  
Pressurized Water ◽  
Leak Testing ◽  
Alternative Methodology ◽  
Water Reactors ◽  
Regulatory Commission

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code [1], Section XI, Appendix G provides a deterministic procedure for defining Service Level A and B pressure–temperature limits for ferritic components in the reactor coolant pressure boundary. An alternative risk-informed methodology has been developed for ASME Section XI, Appendix G. This alternative methodology provides easy to use procedures to define risk-informed pressure–temperature limits for Service Level A and B events, including leak testing and reactor start-up and shut-down. Risk-informed pressure–temperature limits provide more operational flexibility, particularly for reactor pressure vessels with relatively high irradiation levels and radiation sensitive materials. This work evaluated selected plants spanning the population of pressurized water reactors (PWRs) and boiling water reactors (BWRs). The evaluation included determining appropriate material properties, reviewing operating history and system operational constraints, and performing probabilistic fracture mechanics (PFM) analyses. The analysis results were used to define risk-informed pressure–temperature relationships that comply with safety goals defined by the United States (U.S.) Nuclear Regulatory Commission (NRC). This alternative methodology will provide greater operational flexibility, especially for Service Level A and B events that may adversely affect efficient and safe plant operation, such as low-temperature-over-pressurization for PWRs and system leak testing for BWRs. Overall, application of this methodology can result in increased plant efficiency and increased plant and personnel safety.

Effective Approaches for Managing Aging Effects in BWR Reactor Coolant System Components for License Renewal

2003 ◽  
Author(s):  
Barry J. Elliot ◽  
Vikram N. Shah ◽  
Yung Y. Liu
Keyword(s):  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Corrosion Cracking ◽  
Aging Effects ◽  
Irradiation Embrittlement ◽  
Intergranular Stress Corrosion ◽  
Reactor Coolant ◽  
Extended Operation ◽  
License Renewal ◽  
Coolant System

This paper discusses management of aging effects for reactor coolant pressure boundary components in boiling water reactors (BWRs): loss of fracture toughness due to thermal aging and neutron irradiation embrittlement of vessel internals made of cast austenitic stainless steel; cracking of the top guide due to irradiation-assisted stress corrosion cracking; cracking of the core shroud and reactor coolant system piping due to intergranular stress corrosion cracking; cracking of the small bore piping due to high-cycle thermal fatigue; and loss of preload in the pressure boundary bolting. The applicants for license renewal of BWR plants have proposed different approaches for managing these aging effects such that the intended functions of the affected components will be maintained, consistent with the current licensing basis, for the period of extended operation. The NRC staff has performed safety evaluation of these approaches and found them acceptable for adequately managing the aging effects during the period of extended operation. The technical bases for the acceptance are presented in this paper.

Evaluation of the forced outage length contribution of critical nuclear power systems in an extended operating cycle

2000 ◽  
Vol 214 (1) ◽  
pp. 23-28
Author(s):  
C W Kang
Keyword(s):  
Power Systems ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Nuclear Regulatory Commission ◽  
Capacity Factor ◽  
Plant Performance ◽  
Operating Cycle ◽  
Pressurized Water ◽  
Water Reactors ◽  
Regulatory Commission

The work presented here presents an evaluation method for the question of how reliably the system (or component), responsible for the dominant plant availability loss, will run in an extended 48 month operating cycle. As major contributors to the total plant forced outage time in pressurized water reactors (PWRs), reactor coolant pumps (RCPs) and main feed pumps (MFPs) are chosen as specific example systems for a case study. The method proposed estimates the expected forced outage length contribution of each system to the maximum allowed outage length given a certain plant capacity factor. Based upon the current reliability level estimated from the Nuclear Regulatory Commission plant performance database, the assessment of each system impact shows that 14.2 and 2.2 per cent of the maximum allowed outage length are expected to be taken by RCPs and MFPs respectively in the PWR regardless of other systems. In order to meet a 97 per cent goal capacity factor to be envisaged in a 48 month operating cycle, it is recommended that various possible actions be devised for achieving the higher RCP and MFP operational availability through design, monitoring and maintenance.

Verification of New Capabilities of Deterministic Load Module of FAVOR 09.1

2010 ◽  
Author(s):  
Shengjun Yin ◽  
Terry L. Dickson ◽  
Paul T. Williams ◽  
B. Richard Bass
Keyword(s):  
Computational Study ◽  
National Laboratory ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Oak Ridge ◽  
Problem Class ◽  
Wide Range ◽  
Significant Generalization ◽  
Water Reactors ◽  
Regulatory Commission

This paper describes a computational study conducted by the Probabilistic Pressure Boundary Integrity Safety Assessment (PISA) program at Oak Ridge National Laboratory (ORNL) in support of the Nuclear Regulatory Commission (NRC) sponsored verification of the new capabilities of the latest version of Fracture Analysis of Vessels – Oak Ridge (FAVOR) 09.1. The v09.1 version of FAVOR represents a significant generalization over previous versions, because the problem class for FAVOR has been extended to encompass a broader range of transients and vessel geometries. FAVOR, v09.1, provides the capability to perform both deterministic and risk-informed fracture analyses of boiling water reactors (BWRs) as well as pressurized water reactors (PWRs) subjected to heat-up and cool-down transients. In this study, deterministic solutions generated with the FAVOR v09.1 code for a wide range of representative internal/external surface-breaking flaws and embedded flaws subjected to selected thermal-hydraulic transients were benchmarked with the solutions obtained from ABAQUS (version 6.9-1) for the same transients. Based on the benchmarking analyses, it is concluded that the deterministic module implemented into FAVOR, v09.1, satisfies the criteria described in the FAVOR software design documentation.

Stress-Based Environmental Fatigue Monitoring of PWR Charging Nozzle

2013 ◽  
Author(s):  
Timothy Gilman ◽  
Archana Chinthapalli ◽  
Michael Hoehn
Keyword(s):  
Nuclear Power ◽  
Nuclear Regulatory Commission ◽  
Pressurized Water ◽  
Technical Report ◽  
Fatigue Monitoring ◽  
Energy Center ◽  
Plant Data ◽  
Water Reactors ◽  
Fatigue Evaluation ◽  
Regulatory Commission

This paper describes the techniques utilized to perform a stress-based environmentally-assisted fatigue evaluation of Westinghouse-designed charging branch nozzles on the reactor coolant loop of the Callaway Energy Center nuclear power plant. Analysis results from using idealized, design transient definitions are compared to those resulting from analysis of the actual plant data. Benchmarking analyses, performed to address Nuclear Regulatory Commission (NRC) concerns about simplified methodologies, are described. The simplified results are also compared to those produced using an advanced, multiaxial stress-based fatigue methodology defined in a recent EPRI technical report [3]. This paper concludes that stress-based fatigue monitoring using actual plant data is an effective way for a plant to manage environmentally-assisted fatigue of charging nozzles in pressurized water reactors (PWRs).

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